In service conditions, the coolant of a pressurized water reactor (“PWR”) is water that is pressurized to 190 bar and reaches a maximum temperature of 360° C.
In such conditions, the zirconium alloy constituting the nuclear fuel claddings oxidizes in contact with the coolant.
As the oxide formed is brittle, the service life of the claddings is partly limited by the acceptable maximum thickness of oxide. Beyond that, the residual thickness of zirconium alloy that is sound and ductile is insufficient to guarantee good mechanical properties with the aim of ensuring optimal confinement of the nuclear fuel.
Limiting the oxidation in nominal conditions would therefore make it possible to increase the service life of the claddings and therefore improve the burn-up fraction of uranium dioxide.
For this purpose, fuel suppliers have developed new alloys that are more resistant to corrosion in nominal conditions, such as the alloy M5™ from the company Areva-NP.
Although these developments have made it possible to improve the resistance of zirconium alloy claddings to the low-temperature oxidation encountered in nominal conditions, they do not provide decisive gains in the resistance to high-temperature oxidation encountered in accident conditions.
The high temperatures are generally above 700° C., in particular between 800° C. and 1200° C. They are for example reached in the case of hypothetical accident scenarios of the RIA type (Reactivity Insertion Accident) or LOCA type (Loss Of Coolant Accident), or even in conditions of dewatering of the spent fuel storage pool. At such temperatures, the coolant is in the form of steam.
As detailed below, high-temperature oxidation is much more critical than low-temperature oxidation, as deterioration of the cladding, the first barrier for fuel confinement, is quicker and the associated risks are greater. These risks are, among others, as follows:                release of hydrogen;        embrittlement of the cladding at high temperature, by oxidation or even, in certain conditions, hydriding of the cladding;        embrittlement of the cladding on quenching, caused by the sudden temperature drop during the massive supply of water for making the nuclear reactor core safe;        low mechanical strength of the cladding after quenching or cooling, such as in the case of operations for post-accident maintenance.        
Taking these risks into account, it is therefore essential to limit as far as possible the high-temperature oxidation of the cladding in order to improve the safety of nuclear reactors using for instance water as coolant.
Studies aim to improve the oxidation resistance of zirconium alloy subjected to a temperature above that encountered in service conditions. However, the temperatures encountered in accident conditions are not taken into account.
Document “FR 1493040 A” thus proposes coating a zirconium alloy with a layer of chromium. This monolayer coating is presented as making it possible to protect the zirconium alloy against oxidation at a temperature of 600° C., at atmospheric pressure and in carbon dioxide. However, as demonstrated experimentally below, the proposed coating is unable to significantly limit high-temperature oxidation.
It should moreover be noted that the conditions tested in no case allow prediction of the corrosion resistance of the zirconium alloy at a temperature above 600° C.
This lack of predictability is due to the fact that finding solutions against oxidation in accident conditions requires exploring a new field of research. In fact, as detailed below, the physicochemical conditions and the mechanisms of high-temperature oxidation are fundamentally different from those of oxidation at a temperature less than or equal to 600° C.